This invention generally relates to fuel assemblies and more particularly relates to a fuel assembly including deflector vanes for deflecting a component of a fluid stream flowing past such fuel assembly, which fuel assembly may be of the kind typically found in nuclear power reactor cores.
Before discussing the state of the art, it is instructive first to briefly describe the structure and operation of a typical nuclear power reactor, which contains a plurality of nuclear fuel assemblies. In this regard, a nuclear power reactor is a device for producing heat by the controlled fission of nuclear fuel material contained in a plurality of adjacent fuel rods. The fuel rods are bundled together by a plurality of spaced-apart grids, each grid having open cells for receiving each fuel rod therethrough and for obtaining a predetermined distance (i.e., pitch) between the adjacent fuel rods. Moreover, hollow control rod guide thimble tubes are also received through other open cells of each grid. The thimble tubes slidably accept movable absorber or control rods capable of controlling the fission process. A first end portion and a second end portion of each thimble tube are attached to a top nozzle and a bottom nozzle, respectively, for providing rigid structural support to the fuel assembly. The combination of the fuel rods, guide thimble tubes, grids, top nozzle and bottom nozzle is typically referred to in the art as a fuel assembly. A plurality of these fuel assemblies are grouped to define a nuclear reactor pressure vessel.
During operation of the nuclear reactor, a flow stream of liquid moderator coolant (e.g., demineralized water) is caused to flow through the pressure vessel and over the fuel rods for assisting the fission process and for removing the heat produced by fission of the nuclear fuel material contained in each fuel rod. The flow velocity of the coolant, which is pumped over the fuel rods by reactor coolant pumps, may be approximately 18 feet per second, in the case of the typical pressurized water nuclear power reactor, for efficiently removing the heat produced by the fission process. That is, heat due to fission of the nuclear material is transferred from each of the fuel rods, and hence from each fuel assembly, to the liquid moderator coolant flowing past the fuel rods. The heat transferred to the liquid moderator coolant is ultimately carried by the coolant from the pressure vessel to a turbine-generator for generating electricity in a manner well known in the art of electrical power generation. As discussed hereinbelow, it is important for safety reasons that the coolant efficiently removes the heat produced by each fuel rod. For this purpose, the heated surface of each fuel rod should be in contact with the coolant which has a predetermined average bulk coolant temperature.
It is known that the heat flux (i.e., rate of heat transfer per unit area) transversely across the heated surface of the fuel rod will vary as a function of the temperature difference between the heated surface of the fuel rod and the bulk coolant. In order to appreciate the importance of this relationship between heat flux and temperature difference, the discussion immediately hereinbelow provides a description of the manner in which the heat flux varies as a function of the temperature difference between the heated surface of the fuel rod and the bulk coolant. That is, as the difference between the surface temperature of the fuel rod and the bulk coolant is allowed to increase during start-up of the reactor, heat will be transferred from the heated surface to the coolant by single-phase convection, thereby increasing the heat flux. As the difference between the fuel rod surface temperature and the average bulk coolant temperature further increases, the temperature of the heated surface will eventually exceed the saturation temperature (i.e., temperature of saturated steam at the existing pressure in the reactor core) and vapor bubbles will form on the heated surface to produce nucleate boiling on the heated surface in a manner that rapidly increases the heat flux. A maximum heat flux will then occur when the bubbles become dense enough that they coalesce and form a vapor film on the heated surface. However, the vapor film will act as a heat insulator because vapor inhibits heat transfer. This point of maximum heat flux where the vapor film forms on the heated surface is commonly referred to in the art as Departure from Nucleate Boiling (DNB) and is to be avoided for safety reasons. Thus, if the difference between the surface temperature and the bulk temperature is allowed to further increase by even a small amount beyond the maximum heat flux (DNB), the heat flux will rapidly substantially decrease even though the temperature of the heated surface increases. The vapor film on the fuel rod at this point becomes unstable in the sense that the vapor film alternately breaks-down and then reforms so as to produce partial film boiling. If the difference between the surface temperature and the bulk coolant temperature is allowed to increase still further, the heat flux will again increase and stable vapor film boiling will occur. However, if large heat fluxes occur simultaneously with film boiling (i.e., either partial or stable film boiling), the temperature of the heated surface of the fuel rod may become high enough to damage the fuel rod (referred to in the are as "burnout") and is to be avoided for safety reasons. Thus, it is well understood by persons having ordinary skill in the art that if the reactor is operated such that nucleate boiling occurs near DNB, a relatively small increase in the heat flux will cause a relatively rapid change to film boiling that may result in "burnout". Therefore, it is prudent to operate the nuclear reactor such that the highest heat flux is less than the maximum heat flux associated with DNB in order to obtain the highest allowable heat generation without risking damage to the fuel rod.
As discussed hereinabove, a vapor bubble film may form on the heated surface of the fuel rod to produce boiling thereon; however, the vapor film will act as a heat insulator because vapor inhibits heat transfer and may lead to DNB that may in turn lead to fuel rod damage. Hence, it is desirable to maintain a film of liquid substantially single-phase coolant on the surface of the fuel rod to enhance heat transfer from the fuel rod to the coolant while avoiding DNB. Therefore, a problem in the art is to maintain a film of liquid substantially single-phase coolant on the surface of the fuel rod to enhance heat transfer from the fuel rod to the coolant while avoiding DNB.
Enhancing heat transfer from the fuel rod to the coolant while avoiding DNB increases the maximum allowable heat flux obtainable from a given reactor core size. This is desirable because increasing the maximum allowable heat flux obtainable from a given reactor core size increases the maximum allowable power obtainable from the reactor core. In this regard, heat transfer from the fuel rod to the coolant may be enhanced by increasing the bulk coolant flow velocity over the fuel rods. However, increasing the flow velocity of the coolant may require larger and more costly reactor coolant pumps. Therefore, another problem in the art is to more efficiently enhance heat transfer from the fuel rod to the coolant without requiring larger and more costly reactor coolant pumps.
Maintaining a film of liquid substantially single-phase coolant on the surface of the fuel rod to enhance heat transfer from the fuel rod to the coolant while avoiding DNB in a manner not requiring larger coolant pumps has assumed added importance in recent years because some current reactor core designs require the previously mentioned fuel rods to be arranged in a denser triangular pitch array rather than in the more traditional and less dense square pitch array. Thus, in some reactor core designs, the fuel assemblies containing the fuel rods may have a hexagonal transverse cross-section for suitably achieving the "dense-pack" triangular pitch array. Fuel rods arranged in a triangular pitch array obtain a higher average heat flux density from a reactor core of given size compared to fuel rods arranged in the more traditional square pitch array. Obtaining a higher average heat flux density using densely packed fuel assemblies is desirable for economic reasons because such densely packed fuel assemblies achieve more revenue-producing power per unit volume of the reactor core which in turn increases return on plant investment. However, higher heat flux tends to increase the risk of DNB and is therefore undesirable for safety reasons, as discussed hereinabove. Thus, it has become very important to adequately cool such fuel assemblies and the densely packed fuel rods contained therein such that DNB is avoided while simultaneously obtaining a higher heat flux per unit volume of the reactor core.
Fuel assemblies suitable for use in nuclear reactor cores are known. One such fuel assembly is disclosed in U.S. Pat. No. 3,787,285 titled "Fuel Assembly For A Nuclear Reactor And A Nuclear Reactor Core Comprising Such Fuel Assemblies" issued Jan. 22, 1974 in the name of Jorgen Marstrand. This patent discloses a fuel assembly having guide vanes, the axes of which are parallel to the fuel rods and impart a vortical motion to the coolant flowing along the vanes to permit higher energy flux density. The fuel rods are arranged in a hexagonal pattern such that the outer contour of the fuel assembly is hexagonal. A plurality of vanes are disposed about, and tilted with respect to, a central axis to cause the fluid flow over the fuel elements to follow a generally helical path about the central axis. Although the Marstrand patent discloses a fuel assembly having an outer hexagonal contour and a plurality of guide vanes, the Marstrand patent does not appear to disclose a fuel assembly including deflector vanes for deflecting a component of a fluid stream flowing past such fuel assembly, as described and claimed hereinbelow.
Another fuel assembly is disclosed in U.S. Pat. No. 3,281,327 titled "Nuclear Fuel Assemblies" issued Oct. 25, 1966 in the name of John Webb, et al. This patent discloses a spacer grid comprising a support member in the form of an outer metal sleeve of regular hexagonal cross-section. This patent also discloses that the grid has a parallel array of spacer diaphragms adapted to be penetrated by fuel elements and titled with respect to the longitudinal axis of the fuel element. According to this patent, the spacer diaphragms act as deflector vanes imparting to the main flowstream a component of flow transversely of the fuel elements. According to the Webb, et al. patent the diaphragms are advantageous from a heat transfer standpoint because they promote swirling of the coolant to reduce so-called "hot channel factors". Although the Webb, et al. patent discloses a fuel assembly having a parallel array of spacer diaphragms that promote swirling of the coolant to improve heat transfer, the Webb et al. patent does not appear to disclose a fuel assembly including deflector vanes for deflecting a component of a fluid stream flowing past such fuel assembly, as described and claimed hereinbelow.
Although the above recited patents disclose fuel assemblies suitable for use in nuclear reactor cores, these patents do not appear to disclose a fuel assembly including deflector vanes for deflecting a component of a fluid stream flowing past such fuel assembly, as described and claimed hereinbelow.
Therefore, what is needed is a fuel assembly including deflector vanes for suitably deflecting a component of a fluid stream flowing past such fuel assembly.